The present invention relates to a pressurized water reactor plant and, more particularly, the arrangement of various components constituting a primary reactor cooling system therefore.
FIG. 9 shows a typical arrangement of components, such as a reactor vessel and steam generator, constituting a primary cooling system of a conventional pressurized water reactor plant. A high-temperature side pipe 2 of the primary cooling system, which pipe is connected to the outlet nozzle 1a of a reactor vessel 1, communicates with the high-temperature side water chamber 3a of a vertical type steam generator 3. A substantially U-shaped pipe 4, referred to as crossover leg hereinafter, communicating with the low-temperature side water chamber 3b of the steam generator 3 is connected to the suction port 5a of a reactor coolant pump 5. The discharge port 5b of the pump is connected to a low-temperature side pipe (not shown) communicating with the inlet nozzle 1b of the reactor vessel 1.
In such a design that the installations or components are connected as described above, the bottom of the U-shaped pipe (crossover leg) lies below the upper portion of a nuclear core 7 in the reactor vessel 1 as seen from FIG. 9.
The applicant previously proposed a pressurized water reactor plant employing a horizontal type steam generator instead of the vertical type steam generator so that the steam generator can be used not only for generating steam in normal operation but also as a heat exchanger for removing decay heat in a core during a loss of coolant accident (Japanese Patent Application No. 401077/1990, European Patent Application No. 90117428.4)
As seen from FIG. 10, the pressurized water reactor plant disclosed in Japanese Patent Application No. 401077/1990 comprises a condensate tank T1 which is disposed above a secondary cooling system and communicates with the second cooling system via a pipe having a valve, a pressure reducing valve V1 connected to the secondary cooling system, and a gravity flow tank T2 which is partially disposed above a low-temperature side pipe P1 and a high-temperature side pipe P2 of the primary cooling system. Further, the gravity flow tank T2 communicates with a reactor vessel 1 via the pipe P3 provided with a valve V2. Steam generators 3' are installed extending horizontally and have a vent pipe P4 connected to the high-temperature side water chamber. The gravity flow tank T2 has a water capacity sufficient to submerge at least the high-temperature side and low-temperature side pipes P2 and P1 under discharged water when the valve V2, which is disposed in the pipe P3 connected to the gravity flow tank T2, is opened.
In the pressurized water reactor plant mentioned above during a normal reactor operation, the primary coolant which enters the reactor vessel 1 through the low-temperature side pipe P1 connected to the inlet nozzle of the reactor vessel 1 is heated in the nuclear core 7, and flows out of the outlet nozzle of the reactor vessel 1. Then, the primary coolant thus heated passes through the high-temperature side pipe P2, and reaches the low-temperature side water chamber after flowing through the high-temperature side water chamber of the steam generator 3' and the inside of U-shaped heat transfer tubes. Thereafter the primary coolant is pumped by a coolant pump 5, into the reactor vessel 1 again.
While, in the arrangement of the pressurized water reactor plant described above, the natural circulation flow of the reactor coolant through the heat transfer tubes is not prevented (or is sustained), as described in the specification of the above patent application, during a reduced pressure condition in the primary cooling system, which follows from an accident of coolant loss, such that major postulated design basis accident such as core melting by decay heat can be prevented, because the steam generator 3' is installed laterally extending and the high-temperature side water chamber therein is provided with a pipe P4 having a valve. In order to remove the decay heat through heat exchange with a secondary coolant or feedwater in the steam generator 3' by using gravity circulation, the pressure in the secondary cooling line is decreased by operating the pressure reducing valve V1 connected to the secondary cooling system, and the water in the condensate tank T1 is poured into the secondary cooling system under head difference by opening the valve of condensate tank T1 disposed above the secondary cooling systems. Thus the water poured into the secondary cooling system is heat exchanged with the primary coolant flowing through the heat transfer pipe under natural circulation thereby removing the decay heat generated in the nuclear core 7 of the steam generators 3'. Since the steam generator 3' is one of essential components of the primary cooling system, the use of this component eliminates the need for additional equipment, thereby enabling simplification of facility and improvement in economy. Also, since the steam generator 3' is originally designed to remove heat generated in the nuclear core during a normal operation, it has an excellent and sufficient cooling capacity.
Therefore, the invention disclosed in Japanese Patent Application No. 401077/1990 can provide a passive core decay heat removing system which can remove decay heat in a substantially passive manner with the minimum use of active components of high reliability such as valves.
In the improved pressurized water nuclear power plant described above, however, the bottom of the crossover leg, that is the low-temperature side pipe P1 connecting the low-temperature side water chamber of the steam generator 3' and the reactor coolant pump 5, remains below the upper portion of the nuclear core 7 as shown in FIG. 10.
For this reason, in the conventional pressurized water reactor plant, a water stagnant or staying in the crossover leg tends to increase a back pressure in a core when a pipe breaking accident of smaller scale occurs at the outlet side of the reactor coolant pump 5. Such a rise in the back pressure tends to block a supply of water into a nuclear core, when an emergency core cooling water is poured into the primary cooling system, reducing the cooling of the nuclear core.